ML14183A335
| ML14183A335 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/28/1996 |
| From: | Imbro E NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML14183A336 | List: |
| References | |
| NUDOCS 9606100076 | |
| Download: ML14183A335 (11) | |
Text
11-1 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-4001 CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 169 License No. DPR-23
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Carolina Power & Light Company (the licensee), dated January 31, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the.Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.B. of Facility Operating License No. DPR-23 is hereby amended to read as follows:
9606100076 960528 PDR ADOCK 05000261 p
2 B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 169, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION Eugene V. Imbro, Director Project Directorate II-1 Division of Reactor Projects 7 I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 28, 1996
ATTACHMENT TO LICENSE AMENDMENT NO. 169 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages 4.4-1 4.4-1 4.4-2 4.4-2 4.4-3 4.4-3 4.4-4 4.4-4 4.4-8 4.4-8 4.4-9 4.4-9 4.4-10 4.4-10 6.12-1 6.12-1
4.4 CONTAINMENT TESTS Applicability Applies to containment leakage and structural integrity.
Objective To verify that potential leakage from the containment and that pre-stressing tendon loads are maintained within acceptable values.
Speci fi cati on 4.4.1 Operational Leakage Rate Testing Required visual examinations and leakage rate testing shall be performed in accordance with the Containment Leakage Rate Testing Program, except for testing of the containment personnel air lock.
The containment personnel air lock shall be tested every six months.
4.4-1 Amendment No. 101,
THIS PAGE DELETED 4.4-2 Amendment No. 70,
THIS PAGE DELETED 4.4-3 Amendment No.
6F.
4.4.2 Isolation Vave Tests
- a.
Isolation valves shall be tested for operability at each refueling.
- b.
Isolation valves which are pressurized by the penetration pressurization system will be leak tested in accordance with the containment leakage rate testing program.
- c.
The isolation seal water system shall be operated to demonstrate the capability for sealing the associated containment isolation valves at each refueling.
4.4.3 Post Accident Recirculation Heat Removal System
- a.
The portion of the Residual Heat Removal System that is downstream of the first isolation valve outside the containment shall be tested either by use in normal operation or hydrostatically tested at 350 psig at the interval specified below.
- b.
Visual inspection shall be made for excessive leakage from components of the system. Any visual leakage that-cannot be stopped at test conditions shall be measured by collection and weighing or by another equivalent method.
- c.
The acceptance criterion is that maximum allowable leakage from the recirculation heat removal system components (which includes valve stems, flanges and pump seals) shall not exceed two gallons per hour.
- d.
Repairs shall be made as required to maintain leakage with the acceptance criterion in c. above.
4.4-4 Amendment No-
- c.
Notification of the pending test, either of a sample tendon or the containment structural test, along with detailed acceptance criteria shall be forwarded to the Nuclear Regulatory Commission two months prior to the actual test.
Within six months of conducting the test, a report and evaluation shall be submitted to the NRC.
Basis The containment is designed for an accident pressure of 42 psig."' While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a maximum temperature of 1200F.
Post-accident conditions are documented in the Updated Final Safety Analysis Report.
Prior to initial operation, the containment was strength tested at 48.3 psig and then was leak-tested. The acceptance criterion for this preoperational leakage rate test was established as 0.08 weight percent of the contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design pressure of 42 psig. This acceptable leakage rate was equivalent to a 0.1 weight percent of the contained steam-air atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 42 psig and 2630F. The acceptance criteria for Integrated Leakage Rate Tests (ILRTs) is now defined in Technical Specifications Section 6.12. These leakage rates are consistent with the construction of the containment, which is equipped with a penetration pressurization system which pressurizes penetrations, double gasketed seals, and some isolation valve spaces. The channels over all of the containment liner welds were independently leak-tested during construction.
The original safety analysis has been performed on the basis of a leakage rate of 0.10% per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 42 psig and 2630F. With this leakage rate and with minimum containment engineered safety features operating, the public exposure would not exceed 10 CFR 100 guideline values in the event of the design basis accident.")
4.4-8 Amendment No. 78,I c
The performance of a periodic integrated leak rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.
The specified frequency of periodic integrated leak rate tests is based on the following major considerations. First is the low probability of leaks in the liner, because of (a) the test of the leak tightness of the welds during erection: (b) conformance of the complete containment to a low leakage rate limit at the design pressure of 42 psig during preoperational testing which is consistent with 0.1% leakage at design basis accident (DBA) conditions; and (c) absence of any significant stresses in the liner during reactor operation.
4.4-9 Amendment No. IC/
Secondly, the penetration pressurization system is capable of continuously or periodically monitoring leakage from potential leak paths, such as penetrations, double gasketed seals, and spaces between certain containment isolation valves.
Total leakage from the system is measured by summing the recorded flows in each of the four penetration headers. The penetration pressurization system is a qualified system for continuous or intermittent pressurization of individual or groups of containment penetrations as allowed in 10 CFR 50, Appendix J, Items III.B.1.(b), III.B.3.(b), and III.C.1.
A flow sensing device is located in each of the headers supplying make-up air to the four pressurized zones. A leakage rate alarm is provided in each of the four indicating channels to alert the operator in the control room. The flow measurement accuracy is within +/-1%. A flow of 0.04% of the containment volume per day at 42 psig is approximately 0.58 ft'/minute (2.34 scfm).
The flowmeters are capable of indicating leakage well within these limits.
Containment isolation valves are designed to incorporate positive barriers to prevent or minimize leakage through the valves under design basis accident conditions. Several isolation valves are pressurized by the penetration pressurization system to prevent leakage. The remaining valves either receive Isolation Seal Water System water or are installed in systems that are part of a closed system within the containment or operate at system pressures greater than the design pressure of 42 psig in the post-accident condition. These design features provide positive means to prevent containment leakage through the containment isolation valves.
The limiting leakage rates from the recirculation heat removal system are judgment values based primarily on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident. The test pressure, 350 psig, achieved either by normal system operation or hydrostatically testing. gives an adequate margin over the highest pressure within the system after a design basis accident.
4.4-10 Amendment No. 16Q Basis Change, PNSC July 25. 199-
6.12 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50. Appendix J, Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guideline contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program." dated September 1995. Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50.-Appendix J, Option A.
The peak calculated containment internal pressure for the design basis loss of coolant accident Pa is 40 psig.
The maximum allowable containment leak rate, L, at P shall be 0.1% of the containment air weight per day.
a a
Leak Rate acceptance criteria are:
- a.
Containment leakage rate acceptance criteria is less than.or equal to 1.0 L.
During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.6 La for Type B and Type C tests and less than or equal to 0.75 La for Type A tests.
The provisions of Technical Specifications Section 4.0 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.
6.12-1 Amendment No.169